Ring tensile tests of a high burn-up nuclear fuel cladding
Creators
Description
Related to the degradation of the mechanical properties of Zr-based nuclear fuel cladding tubes under severe accident conditions such as RIA(Reactivity Initiated Accident) and LOCA(Loss of Coolant Accident), the mechanical tests which can simulate the degradation of the mechanical properties properly are needed. Therefore, in this paper, modified ring tensile tests were performed in order to evaluate the mechanical properties of high burn-up fuel cladding under a hoop loading condition in a hot cell. The mechanical properties comprise of the strength such as the YS(yield strength) and the UTS(Ultimate Tensile Strength) as well as the ductility such as the UE(Uniform Elongation) and the TE(Total Elongation). The tests were performed with Zircaloy-4 nuclear fuel cladding whose burn-up is approximately 57,000 MWd/tU in the temperature range of room temperature to 800 C. All the experiments were carried out at a constant strain rate of 0.01/s. As the first step, the hoop loading grip for the ring tensile test was designed in order that a constant curvature of the specimen was maintained during the deformation, and the graphite lubricant was used to minimize the friction between the outer surface of the die insert and the inner surface of the ring specimen. The specimen for the ring tensile test was designed to limit the deformation within the gauge section and to maximize the uniformity of the strain distribution. It was confirmed that the mechanical properties under a hoop loading condition can be correctly evaluated by using this test technique. On the basis of the ring tensile tests for a high burn-up Zircaloy-4 cladding, the following conclusions were drawn. Firstly, the mechanical properties are abruptly degraded beyond 600C, which corresponds to a design basis accident condition such as RIA. Secondly, the high burn-up Zircaloy-4 cladding showed a brittle fracture behavior even at the high temperatures (e.g. over 600C) which are achievable during a RIA. (Author)
Additional details
Publishing Information
- Imprint Title
- Proceedings of the fuel and materials sessions
- Imprint Pagination
- 175 p.
- Journal Page Range
- p. 77-84
- Report number
- HPR--364-V1
Conference
- Title
- Enlarged Halden programme group meeting
- Dates
- 16-21 Oct 2005
- Place
- Lillehammer (Norway)
INIS
- Country of Publication
- Norway
- Country of Input or Organization
- Norway
- INIS RN
- 43106748
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S11: NUCLEAR FUEL CYCLE AND FUEL MATERIALS;
- Resource subtype / Literary indicator
- Non-conventional Literature, Conference
- Descriptors DEI
- CLADDING; FUEL ELEMENT FAILURE; FUEL-CLADDING INTERACTIONS; FUEL-COOLANT INTERACTIONS; LOSS OF COOLANT; MEETINGS; NUCLEAR FUELS; REACTOR ACCIDENT SIMULATION; REACTOR ACCIDENTS; REACTOR SAFETY; SYSTEM FAILURE ANALYSIS; TENSILE PROPERTIES; ZIRCALOY 4
- Descriptors DEC
- ACCIDENTS; ALLOYS; ALLOY-ZR98SN-4; CHROMIUM ADDITIONS; CHROMIUM ALLOYS; CORROSION RESISTANT ALLOYS; DEPOSITION; ENERGY SOURCES; FAILURES; FUELS; HEAT RESISTANT MATERIALS; HEAT RESISTING ALLOYS; IRON ADDITIONS; IRON ALLOYS; MATERIALS; MECHANICAL PROPERTIES; REACTOR ACCIDENTS; REACTOR MATERIALS; SAFETY; SIMULATION; SURFACE COATING; SYSTEMS ANALYSIS; TIN ALLOYS; TRANSITION ELEMENT ALLOYS; ZIRCALOY; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS